[Federal Register Volume 68, Number 52 (Tuesday, March 18, 2003)]
[Notices]
[Pages 12946-12964]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-6286]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from, February 21, 2003, through March 6, 2003.
The last biweekly notice was published on March 4, 2003 (68 FR 10277).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission
[[Page 12947]]
expects that the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administrative Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the Commission's Public
Document Room (PDR), located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
By April 17, 2003, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR part 2. Interested
persons should consult a current copy of 10 CFR 2.714,\1\ which is
available at the Commission's PDR, located at One White Flint North,
Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville,
Maryland. Publicly available records will be accessible from the
Agencywide Documents Access and Management System's (ADAMS) Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing
or petition for leave to intervene is filed by the above date, the
Commission or an Atomic Safety and Licensing Board, designated by the
Commission or by the Chairman of the Atomic Safety and Licensing Board
Panel, will rule on the request and/or petition; and the Secretary or
the designated Atomic Safety and Licensing Board will issue a notice of
a hearing or an appropriate order.
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\1\ The most recent version of Title 10 of the Code of Federal
Regulations, published January 1, 2002, inadvertently omitted the
last sentence of 10 CFR 2.714 (d) and paragraphs (d)(1) and (d)(2)
regarding petitions to intervene and contentions. For the complete,
corrected text of 10 CFR 2.714 (d), please see 67 FR 20884; April
29, 2002.
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As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemaking and
Adjudications Staff, or may be delivered to the Commission's PDR,
located at One White Flint North, Public File Area 01F21, 11555
Rockville Pike (first floor), Rockville, Maryland, by the above date.
Because of continuing disruptions in delivery of mail to United States
Government offices, it is requested that petitions for leave to
intervene and requests for hearing be transmitted to the Secretary of
the Commission either by means of facsimile transmission to 301-415-
1101 or by e-mail to [email protected]. A copy of the request for
hearing and petition for leave to intervene should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and because of continuing disruptions in
delivery of mail to United States Government offices, it is requested
that copies be transmitted either by means of facsimile transmission to
301-415-3725 or by e-mail to [email protected]. A copy of the
request for hearing and petition for leave to intervene should also be
sent to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
[[Page 12948]]
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there
are problems in accessing the documents located in ADAMS, contact the
NRC PDR Reference staff at 1-800-397-4209, 301-415-4737 or by e-mail to
[email protected].
AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania
Date of amendment request: January 16, 2003
Description of amendment request: The proposed amendment would
revise the TMI-1 Technical Specifications to incorporate changes
associated with the Cycle 15 core reload design analysis.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed Technical Specification limits (Figure 2.1-1) and
reactor protection system (RPS) trip setpoints (Table 2.3-1) are
developed in accordance with the methods and assumptions described
in NRC-[Nuclear Regulatory Commission] approved Framatome ANP
Topical Reports BAW-10179 P-A, ``Safety Criteria and Methodology for
Acceptable Cycle Reload Analyses'' and BAW-10187 P-A, ``Statistical
Core Design for B&W-[Babcock&Wilcox-] Designed 177 FA Plants.'' The
core thermal-hydraulic code (LYNXT) and CHF [critical heat flux]
correlation (BWC) have been approved for use with these methods and
the Mark-B fuel type utilized at TMI Unit 1. The proposed Technical
Specification requirements on Variable Low Pressure Trip (VLPT)
instrument operating conditions (Table 3.5-1) and surveillances
(Table 4.1-1) are consistent with the VLPT requirements that were
last contained in the TMI Unit 1 Technical Specifications prior to
Cycle 7. The existing flux-flow trip setpoint and power/pump monitor
trip have been shown to provide adequate DNB [departure from
nucleate boiling] protection for Updated Final Safety Analysis
Report (UFSAR) DNB-limiting loss of coolant events.
The margin retained for penalties such as transition core
effects, by imposing a Thermal Design Limit of 1.40 in all DNB
analyses supporting the proposed change, has been shown to be
sufficient to offset the current mixed core conditions at TMI Unit
1, where the Mark-B12 fuel design with fine mesh debris filter is
co-resident with earlier, non-debris filter Mark-B fuel designs.
Therefore the previous commitment to require a higher minimum RCS
[reactor coolant system] flow (105.5% of design flow instead of
104.5%) to offset transition core penalties is no longer necessary.
Reload cycles are designed and operated with maximum steady-
state radial-local peaking factors that are bounded by UFSAR
assumptions used to determine the dose consequences from fuel
handling accidents.
The proposed change to Technical Specification 3.5.2.2.a is only
an administrative correction.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed Technical Specification limits (Figure 2.1-1) and
reactor protection system (RPS) trip setpoints (Table 2.3-1) provide
core protection safety limits and Variable Low Pressure Trip
setpoints developed in accordance with NRC-approved methods and
assumptions. The transition core penalty resulting from Mark-B12
fuel with fine mesh debris filters co-residing with earlier, non
debris filter Mark-B fuel has been demonstrated to be sufficiently
bounded by the analyses supporting the proposed amendment. Therefore
the previous commitment to require a higher minimum RCS flow (105.5%
of design flow instead of 104.5%) to offset transition core
penalties is no longer necessary. These changes have been evaluated
for their impact on the design and operation of plant structures,
systems, and components. These changes do not introduce any new
accident precursors and do not involve any alterations to plant
configurations, which could initiate a new or different kind of
accident.
The proposed change to Technical Specification 3.5.2.2.a is only
an administrative correction.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed reactor protection system (RPS) trip setpoints
(Table 2.3-1) ensure core protection safety limits will be preserved
during power operation. The proposed safety limits and setpoints are
developed in accordance with NRC-approved methods and assumptions.
The margin retained for penalties such as transition core effects,
by imposing a Thermal Design Limit of 1.40 in all DNB analyses
supporting the proposed change, has been shown to be sufficient to
offset the current mixed core conditions at TMI Unit 1. The margin
available between minimum DNBR [departure from nucleate boiling
ratio] results for UFSAR loss of coolant flow events and the Thermal
Design Limit of 1.40 is significant and is similar to DNB margin
results for the current non-SCD [Statistical Core Design] analysis.
Reload cycles are designed and operated with maximum steady-
state radial-local peaking factors that are bounded by UFSAR
assumptions used to determine the dose consequences from fuel
handling accidents.
The proposed change to Technical Specification 3.5.2.2.a is only
an administrative correction.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Edward J. Cullen, Jr., Esquire, Vice
President, General Counsel and Secretary, Exelon Generation Company,
LLC, 300 Exelon Way, Kennett Square, PA 19348.
NRC Section Chief: Richard J. Laufer.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of amendment request: February 14, 2003.
[[Page 12949]]
Description of amendment request: The amendment would allow an
increase in the maximum decay heat of spent fuel stored in Spent Fuel
Pools (SFPs) C and D from 1.0 MBTU/hr to 7.0 MBTU/hr in Technical
Specification 5.6.3.d. The amendment would also increase the allowable
SFP temperatures from 140 degrees F to 150 degrees F under normal and
emergency conditions other than a design-basis Loss-of-Coolant Accident
(LOCA). For a LOCA, the maximum allowed SFP temperature would increase
from 150 degrees F to 160 degrees F.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
A written evaluation of the significant hazards consideration of
a proposed license amendment is required by 10 CFR 50.92. Progress
Energy Carolinas, Inc. (alternately known as Carolina Power & Light
Company) has evaluated the proposed amendment and determined that it
involves no significant hazards consideration. According to 10 CFR
50.92, a proposed amendment to an operating license involves no
significant hazards consideration if operation of the facility in
accordance with the proposed amendment would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated; or
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated; or
3. Involve a significant reduction in a margin of safety.
The basis for this determination is as follows:
Proposed Change
The change involves an increase in the maximum decay heat of
spent fuel stored in Spent Fuel Pools (SFPs) C and D from 1.0 MBTU/
hr to 7.0 MBTU/hr, and an increase in the allowable SFP
temperatures.
Basis
This change does not involve a significant hazards consideration
for the following reasons:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The license amendment only increases the heat load from the Fuel
Pool Cooling and Cleanup System (FPCCS) and the maximum allowable
pool temperature. The changes do not modify the design of
Structures, Systems and Components (SSCs) that could initiate an
accident. The FHB [Fuel Handling Building] Emergency Exhaust System
mitigates the consequences of a fuel handling accident in the Fuel
Handling Building. This system has been evaluated for the conditions
that would exist with the higher SFP temperatures and it was found
that there would be no decrease in the charcoal efficiency. As a
result, there was no increase in the doses from the fuel handling
accident in the FHB. Therefore, the change does not result in any
increase in the probability or consequences in any accident
previously analyzed.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The increase in the SFP decay heat load and the SFP temperature
limit does not involve new plant components or procedures. No
significant impact on any postulated accident is made due to this
change since the required cooling capacity is maintained to the SFPs
and the FPCCS, and the SFPs will operate within design parameters.
For the activation of SFPs C and D, Progress Energy Carolinas,
Inc. performed a Probabilistic Safety Analysis (PSA) of a total loss
of SFP forced cooling. That analysis concluded that the probability
of spent fuel rack uncovery was not credible. That analysis remains
bounding for this license amendment application.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed amendment does not involve a significant
reduction in the margin of safety.
The proposed changes do not affect the design or operation of
the barriers to fission product release (fuel cladding, reactor
coolant system pressure boundary, and containment boundary). The
change in the SFPs C and D decay heat load is bounded by the heat
load used in the analysis of the safety-related systems for design
basis accidents. Therefore, there is no impact in the margin of
safety.
Based on these considerations, the proposed change does not
involve a significant reduction on the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: William D. Johnson, Vice President and
Corporate Secretary, Carolina Power & Light Company, Post Office Box
1551, Raleigh, North Carolina 27602.
NRC Section Chief: Allen Howe.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of amendment request: February 19, 2003.
Description of amendment request: The proposed amendments would
revise Technical Specification (TS) 5.5.10, ``Steam Generator (SGs)
Tube Surveillance Program.'' The proposed amendments would relocate to
TS 5.5.21 the TS 5.5.10 program requirements that apply to the original
SGs and would provide a new TS 5.5.10 that contains program
requirements that would apply to the new SGs when they are installed.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Pursuant to 10 CFR 50.91, Duke has made the determination that
this amendment request does not involves a significant hazard by
applying the three standards established by the NRC regulations in
10 CFR 50.92 as described below.
First Standard
The proposed amendment would not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed amendment will revise Technical Specification (TS)
5.5.10 to delete and clarify replacement steam generator (SG)
surveillance requirements applicable to the replacement of the SGs
following their installation. The proposed amendment does not result
in any changes to the design or methods of operation of the facility
or any of its structures, systems or components (SSC). The SG repair
methods that would be deleted are not applicable to the replacement
SGs due to the use of improved materials and design. Defects found
during future replacement SG tube inspections that exceed the limits
in the new TS 5.5.10 will be removed from service by plugging rather
than being repaired. The accident analyses and assumptions made in
the Updated Final Safety Analysis Report (UFSAR) Chapter 15,
Accident Analyses, are not changed as a result of the proposed
changes. There are no changes resulting from the new TS 5.5.10 that
could affect the function of preventing or mitigating any of these
accidents. The proposed change does not increase the likelihood of
the malfunction of an SSC that may increase the probability or
consequences of an accident. The relocated surveillance requirements
for the current steam generators will not change as a result of the
proposed TS changes. Therefore, the proposed change will not result
in a significant increase in the probability or consequences of an
accident previously evaluated.
Second Standard
The proposed amendment would not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes to the SG tube surveillance TS will delete
or modify surveillance requirements that would otherwise not be
applicable to the replacement steam generators. SG Tubes found to
exceed the plugging limit criteria of TS 5.5.10 for continued
[[Page 12950]]
service will be removed from service by plugging rather than being
repaired. The plugging limit is unchanged by the proposed amendment.
These changes will not introduce any adverse changes to the
facilities' design bases or postulated accidents resulting from
potential tube degradation. The proposed amendment does not affect
the design of SGs, their method of operation, or primary coolant
chemistry controls. In addition, the proposed amendment does not
impact any other SSC. Surveillance requirements for the current SGs
will not change prior to their removal from service as a result of
the proposed changes. Therefore, the proposed changes do not create
the possibility of a new or different type of accident from any
accident previously evaluated.
Third Standard
The proposed amendment would not involve a significant reduction
in the margin of safety.
Margin of safety is related to the confidence in the ability of
the fission product barriers to perform their design functions
during and following an accident situation. These barriers include
the fuel cladding, the reactor coolant system, and the containment
system. These barriers are unaffected by the changes proposed in
this LAR. The steam generator tubes are an integral part of the
reactor coolant pressure boundary. Repairing SG tubes by previously
approved methods of sleeving or rerolling are considered to be an
equivalent boundary to plugging a steam generator tube as has also
been previously approved. Therefore, the margin of safety is not
reduced by the changes proposed in this license amendment request.
Conclusion
Based upon the proceeding evaluation, performed pursuant to 10
CFR 50.92, Duke Energy Corporation has concluded that approval and
implementation of this license amendment request at the Oconee
Nuclear Station will not involve a significant hazards
consideration. The proposed changes revise the steam generator
surveillance requirements to be consistent with the replacement
steam generators. Following implementation of the changes proposed
in this license amendment request, the Oconee steam generators will
continue to be operated in a safe and conservative manner.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Anne W. Cottington, Winston and Strawn, 1200
17th Street, NW., Washington, DC 20005.
NRC Section Chief: John A. Nakoski.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of amendment request: January 29, 2003.
Description of amendment request: The proposed amendment would
change the spent fuel pool loading restrictions by redefining the
regions, inserting Metamic[reg] poison panels in a portion of the spent
fuel pool, and increasing the minimum boron concentration.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
Most accident conditions will not result in an increase in K-
effective (Keff) of the fuel stored in the rack. However,
there are accidents that can be postulated to increase reactivity.
For these accident conditions, the double contingency principle of
ANS [American Nuclear Society] N16.1-1975 is applied. This states
that it is unnecessary to assume two unlikely, independent,
concurrent events to ensure protection against a criticality
accident. Therefore, for accident conditions, the presence of
soluble boron in the storage pool water can be assumed as a
realistic initial condition since its absence would be a second
unlikely event.
A vertical drop accident condition directly upon a cell will
cause damage to the racks in the active fuel region. The proposed
2000 ppm [parts per million] TS [technical specification]
limit will insure that Keff does not exceed 0.95. A fuel
assembly dropped on top of the rack will not deform the rack
structure such that criticality assumptions are invalidated. The
rack structure is such that [after rack deformation] an assembly
positioned horizontally on top of the rack is more than eight inches
away from the upper end of the active fuel region of the stored
assemblies. This distance precludes interaction between the dropped
assembly and the stored fuel. An inadvertent drop of an assembly
between the outside periphery of the rack and the pool wall is
bounded by the worst case fuel misplacement accident condition of
825 ppm. The distance between all the rack modules and the pool
walls is [nominally] less than the width of a fuel assembly.
The fuel assembly misplacement accident was considered for all
storage configurations. An assembly with high reactivity is assumed
to be placed in a storage location which requires restricted storage
based on initial U-235 [Uranium-235] loading and burnup. The
presence of boron in the pool water assumed in the analysis has been
shown to substantially offset the worst case reactivity effect of a
misplaced fuel assembly for any configuration. The boron requirement
of 825 ppm is less than the proposed 2000 ppm minimum
boron TS limit. Therefore, a five percent subcriticality margin can
be easily met for postulated accidents since any reactivity increase
will be much less than the negative worth of the dissolved boron.
For fuel storage applications, water is present. An ``optimum
moderation'' accident is not a concern in spent fuel pool storage
racks because the rack design prevents the preferential reduction of
water density between the cells of a rack (e.g., boiling between
cells).
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes will define a portion of the current Region
2 as Region 3. The new region will contain Metamic[reg] poison panel
inserts and will allow unrestricted storage of fuel assemblies with
various enrichments and burnup. To support the proposed change, a
new criticality analysis was performed. The analysis resulted in new
loading restrictions in Region 1 and Region 2. The presence of boron
in the pool water assumed in the analysis is less than the proposed
ANO-2 [Arkansas Nuclear One, Unit 2] TS minimum concentration of
2000 ppm. Therefore, a five percent subcriticality margin
can be easily met for postulated accidents since any reactivity
increase will be much less than the negative worth of the dissolved
boron.
No new or different types of fuel assembly drop scenarios are
created by the proposed change. During the installation of the
Metamic[reg] panels, the possible drop of a panel is bounded by the
current fuel assembly drop analysis. No new or different fuel
assembly misplacement accidents will be created. Administrative
controls currently exist to assist in assuring that fuel
misplacement does not occur.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
With the presence of a nominal boron concentration, the SFP
[spent fuel pool] storage racks are designed to assure that fuel
assemblies of less than or equal to five weight percent U-235
enrichment when loaded in accordance with the proposed loading
restrictions will be maintained within a subcritical array with a
subcritical margin of
[[Page 12951]]
five percent. This has been verified by criticality analyses.
Credit for soluble boron in the SFP water is permitted under
accident conditions. The proposed change that will allow insertion
of Metamic[reg] poison panels does not result in the potential of
any new misplacement scenarios. Criticality analyses have been
performed to determine the required boron concentration that would
ensure that the maximum Keff does not exceed 0.95. By
increasing the minimum boron concentration to 2000 ppm,
the margin of safety currently defined by taking credit for soluble
boron will be maintained.
The structural analysis of the spent fuel racks along with the
evaluation of the SFP structure showed that the integrity of these
structures will be maintained with the addition of the poison
inserts. All structural requirements were shown to be satisfied, so
all the safety margins were maintained.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point
Nuclear Generating Unit No. 3, Westchester County, New York
Date of amendment request: December 17, 2002.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 5.5.10, ``Ventilation Filter
Testing Program,'' to adopt the requirements of the American Society
for Testing and Materials Standard (ASTM) D3803-1989, ``Standard Test
Method for Nuclear-Grade Activated Carbon.'' The proposed TS revisions
are in response to Nuclear Regulatory Commission (NRC) Generic Letter
(GL) 99-02, ``Laboratory Testing of Nuclear-Grade Activated Charcoal.''
The NRC had previously published a notice of consideration on December
12, 2001 (66 FR 64292) regarding a similar proposal from the licensee
in response to GL 99-02. However, in response to a request for
additional information from the NRC dated March 29, 2002, the licensee
has now revised its proposed amendment. In addition to withdrawing the
prior request to change the maximum control room ventilation system
(CRVS) differential pressure in TS 5.5.10.d, the proposed amendment
would revise the TSs: (1) To provide a CRVS methyl iodide removal
efficiency of greater than or equal to 95.5% and remove the notation
that there is a 1-inch charcoal bed depth; (2) to allow for the
continued use of the existing CRVS through Refueling Outage 13, in
order to design, fabricate, and install a 2-inch charcoal filter bed;
(3) to add a note in the TS requiring a demonstration of charcoal
efficiency of 93% when changing the charcoal in the existing CRVS bed
prior to any fuel movement in the upcoming Refueling Outage 12 and
every 6 months thereafter until the new beds are installed. The
proposed amendment also seeks an exception from the factor of safety of
two for the Containment Fan Cooler Units due to the plant's design.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
Response: The proposed license amendment adopts the new test
method and acceptance criteria of ASTM D3803-1989 for activated
charcoal filters. The changes require laboratory performance testing
of adsorber carbon that yields a more accurate result than the
testing currently required by the TS. The proposed change to delete
non-conservative TS requirements for testing of adsorber carbon is
not a plant accident initiator as described in the Final Safety
Analysis Report (FSAR). The proposed amendment does not change the
function of any structure, system or component (SSC). The function
of the ventilation systems is filtration of radiological releases
during postulated accidents. The proposed changes will provide
greater assurance that this function is provided. The revised TS
requirements are for laboratory tests that are currently in place to
address Generic Letter 99-02, with one exception to the safety
factor of 2, and accommodate the change of the Control Room
Ventilation System (CRVS) charcoal beds to two inches. The change
only affects the TS testing requirements since the modification to
the CRVS will be accomplished separately from the TS change. The TS
changes will not result in any changes to the efficiency assumed in
accident analysis. The changes do not alter, degrade or prevent
actions described or assumed in an accident described in the FSAR.
Therefore, the proposed amendment does not change the possibility of
an accident previously evaluated or significantly increase the
consequences of an accident previously evaluated.
(2) Does the proposed license amendment create the possibility
of a new or different kind of accident from any accident previously
evaluated?
Response: The proposed license amendment adopts the new test
method and acceptance criteria of ASTM D3803-1989 for activated
charcoal filters. The change does not involve any modifications to
the plant but will accommodate the planned modification of the CRVS
to change the charcoal beds from 1 inch to 2 inches. The change will
not require changes to how the plant is operated nor will it affect
the operation of the plant. The changes require laboratory
performance testing of adsorber carbon that yields a more accurate
result than the testing currently required by the TS. The proposed
changes to delete non-conservative TS requirements for testing of
adsorber carbon is not a plant accident initiator as described in
the Final Safety Analysis Report (FSAR). The proposed amendment does
not change the function of any structure, system or component (SSC).
The function of the ventilation systems is filtration of
radiological releases during postulated accidents. The proposed
changes will provide greater assurance that this function is
provided. The revised TS requirements are for laboratory tests that
are currently in place to address Generic Letter 99-02, with one
exception to the safety factor of 2, and accommodate the change of
the Control Room Ventilation System (CRVS) charcoal beds to two
inches. The change only affects the TS testing requirements since
the modification to the CRVS will be accomplished separately from
the TS change. The TS changes will not result in any changes to the
efficiency assumed in accident analysis. The changes do not alter,
degrade or prevent actions described or assumed in an accident
described in the FSAR. Therefore, the proposed amendment does not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
(3) Does the proposed license amendment involve a significant
reduction in a margin of safety?
Response: The proposed license amendment adopts the new test
method and acceptance criteria of ASTM D3803-1989 for activated
charcoal filters. The proposed license amendment does not reduce the
margin of safety but enhances it by requiring more accurate testing.
The proposed test change will require the use of a current and
improved ASTM standard to ensure that the carbon ability to adsorb
radioactive material will remain at or above the capability credited
in our accident analysis.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. John Fulton, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY
10601.
NRC Section Chief: Richard J. Laufer.
[[Page 12952]]
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of amendment request: January 23, 2003.
Description of amendment request: The proposed amendment would
modify the Pilgrim Nuclear Power Station Technical Specification (TS)
requirements for the Emergency Core Cooling System (ECCS) during
shutdown conditions. The proposed amendment would change the Core Spray
and Low Pressure Coolant Injection System's TS requirements to be
applicable during the Run, Startup, and Hot Shutdown Modes. The
proposed change would also modify the High Drywell Pressure
Instrumentation TSs to require the instrumentation to be Operable
during the Run, Startup and Hot Shutdown Modes. The proposed change
would also remove unnecessary TS requirements based on the plant's
operating Mode. Other proposed changes are administrative in nature.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's review is
presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
The proposed change involves modifications to the TS operability
requirements for the ECCS during shutdown conditions. The ECCS is
designed to mitigate the release of radioactive materials to the
environment following a Loss of Coolant Accident (LOCA). The
modifications remove certain ECCS TS requirements during shutdown
conditions and includes additional requirements for the Cold
Shutdown or Refuel Modes when the availability of the ECCS is most
likely to be needed. The additional requirements are more
restrictive and are proposed to reduce the probability or
consequences of potential accidents. The requirements proposed to be
removed are unnecessary due to the associated plant conditions and
other changes are administrative in nature. No increase in the
probability or consequences of an accident previously evaluated has
been identified for these changes. The ECCS is not an initiator of
any accidents previously evaluated and the proposed change does not
increase the amount of radioactive materials available to be
released for a previously evaluated accident. Therefore, the
proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed change involves modifications to the TS operability
requirements for the ECCS during shutdown conditions. The
modifications remove unnecessary ECCS TS requirements during
shutdown conditions and includes additional requirements for the
Cold Shutdown or Refuel Modes when the availability of the ECCS is
most likely to be needed. In addition, the proposed change makes
administrative changes. The proposed change does not involve any
physical alteration of ECCS equipment and does not create a new mode
of system operation. In addition, no new or different types of ECCS
equipment will be installed as a result of the proposed change. The
proposed change will allow the installation of modifications on the
reference and variable legs of the instrument racks that support the
ECCS and Feedwater level instrumentation. No other types of
accidents or accident initiators associated with the proposed change
or modifications have been identified. Therefore, the proposed
change does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
The ECCS is designed to mitigate the release of radioactive
materials to the environment following a LOCA. The long-term cooling
analysis following a design basis LOCA demonstrates that only one
low-pressure ECCS injection/spray subsystem is required, post LOCA,
to maintain adequate reactor vessel water level. The proposed change
includes an additional requirement that two low-pressure injection/
spray subsystems be Operable for the Cold Shutdown or Refuel Modes.
The requirements proposed to be removed are unnecessary due to the
associated plant conditions and other proposed changes are
administrative in nature. No scenario has been identified that, as a
result of the proposed change, would create a single component
failure which prevents the automatic initiation of the ECCS. The
proposed change will not modify the method by which any safety-
related system performs its function and ECCS operation and testing
will remain consistent with current safety analysis assumptions.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: J. M. Fulton, Esquire, Assistant General
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth,
Massachusetts 02360-5599.
NRC Section Chief: James W. Clifford.
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Date of amendment request: December 20, 2002.
Description of amendment request: The proposed amendments would
remove technical specification requirements for reactor protection
system Function 5, main steam isolation valve closure, and Function 10,
turbine condenser vacuum low, when in startup.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed changes to the Dresden Nuclear Power Station (DNPS)
Units 2 and 3 Technical Specifications (TS) revise the applicability
of TS 3.3.1.1, ``Reactor Protection System (RPS) Instrumentation,''
Function 5 (i.e., Main Steam Isolation Valve--Closure) and Function
10 (i.e., Turbine Condenser Vacuum--Low) to eliminate the
requirement for these functions to be operable while in Mode 2 with
reactor pressure =600 psig. The proposed changes also
delete Required Action F.2 of TS 3.3.1.1 to align with the revised
applicability for Functions 5 and 10.
TS requirements that govern operability or routine testing of
plant instruments are not assumed to be initiators of any analyzed
event because these instruments are intended to prevent, detect, or
mitigate accidents. Therefore, these proposed changes will not
involve an increase in the probability of an accident previously
evaluated.
Additionally, these proposed changes will not increase the
consequences of an accident previously evaluated because the
proposed changes do not adversely impact structures, systems, or
components. These changes will not alter the operation of equipment
assumed to be available for the mitigation of accidents or
transients by the plant safety analysis. Functions 5 and 10 are
currently required in Mode 2 with reactor pressure =600
psig to ensure that the reactor is shut down to prevent an
overpressurization transient due to closure of main steam isolation
valves or turbine stop valves. The existing scram logic is the
result of experience gained during the startup of an early vintage
boiling water reactor in 1966 when operators had difficulty
controlling reactor power above approximately 600 psig without
pressure control. Experience on later plant startups indicates that
the early experience may not be inherent to the boiling water
reactor design. As such, General Electric subsequently recommended
that the scram requirement be eliminated. In Mode 2, the heat
generation rate is low enough so that the
[[Page 12953]]
other diverse RPS functions provide sufficient protection from an
overpressurization transient. Furthermore, there will be no change
in the types or significant increase in the amounts of any effluents
released offsite.
For these reasons, the proposed changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes revise the applicability for Functions 5
and 10 of TS 3.3.1.1. The RPS is not an initiator of any accident.
Rather, the RPS is designed to initiate a reactor scram when one or
more monitored parameters exceed their specified limits to preserve
the integrity of the fuel cladding and the reactor coolant pressure
boundary and minimize the energy that must be absorbed following an
accident. The proposed changes do not alter the applicability for
RPS functions during plant conditions in which an overpressurization
transient is assumed to occur. Therefore, the proposed changes do
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
Margins of safety are established in the design of components,
the configuration of components to meet certain performance
parameters, and in the establishment of setpoints to initiate alarms
and actions. The proposed changes revise the applicability for
Functions 5 and 10 of TS 3.3.1.1. The proposed changes do not alter
the applicability for RPS functions during plant conditions in which
an overpressurization transient is assumed to occur. In addition,
the proposed changes do not affect the probability of failure or
availability of the affected instrumentation. Furthermore, the
proposed changes will reduce the probability of test-induced plant
transients and equipment failures. Therefore, the proposed changes
do not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Edward J. Cullen, Deputy General
Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
NRC Section Chief: Anthony J. Mendiola.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412,
Beaver Valley Power Station, Unit 2, Beaver County, Pennsylvania
Date of amendment request: February 4, 2003.
Description of amendment request: The proposed amendment would
extend the surveillance interval of the slave relay in the Engineered
Safety Feature Actuation System instrumentation from 92 days to 12
months. The proposed amendment includes changes to surveillance
requirement (SR) 4.3.2.1.1 and the related Bases.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No. The proposed change to the slave relay test
interval reduces the potential for spurious actuation of equipment,
and therefore does not increase the probability of any accident
previously analyzed. The proposed change to the slave relay test
interval does not change the response of the unit to any accidents
and has an insignificant impact on the reliability of the engineered
safety feature actuation system (ESFAS) signals. The ESFAS will
remain highly reliable and the proposed change will not result in a
significant increase in the risk of plant operation. This is
demonstrated by showing that the impact on plant safety as measured
by the change in core damage frequency (CDF) is less than 1.0E-06
per year and the change in large early release frequency (LERF) is
less than 1.0E-07 per year. The change meets the acceptance criteria
in Regulatory Guide 1.174. Therefore, since the ESFAS will continue
to perform its function with high reliability as originally assumed,
and the increase in risk as measured by the change in CDF and LERF
is within the acceptance criteria of existing regulatory guidance,
there will not be a significant increase in the consequences of any
accidents.
The proposed change does not adversely affect accident
initiators or precursors nor alter the design assumptions,
conditions, or configuration of the facility or the manner in which
the unit is operated and maintained. The proposed change does not
alter or prevent the ability of structures, systems, and components
(SSCs) from performing their intended function to mitigate the
consequences of an initiating event within the assumed acceptance
limits. The proposed change does not affect the source term,
containment isolation, or radiological release assumptions used in
evaluating the radiological consequences of an accident previously
evaluated. Further, the proposed change does not increase the types
or amounts of radioactive effluent that may be released offsite, nor
significantly increase individual or cumulative occupational/public
radiation exposures. The proposed change is consistent with the
safety analysis assumptions and resultant consequences.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No. The proposed change does not result in a change in
the manner in which the EFSAS provides unit protection. The EFSAS
will continue to have the same setpoints after the proposed change
is implemented. There are no design changes associated with the
proposed change. The change to the slave relay test interval does
not change any existing accident scenarios, nor create any new or
different accident scenarios.
The change does not involve a physical alteration to the unit
(i.e., no new or different type of equipment will be installed) or a
change in the methods governing normal plant operation. In addition,
the change does not impose any new or different requirements or
eliminate any existing requirements. The change does not alter
assumptions made in the safety analysis. The proposed change is
consistent with the safety analysis assumptions and current unit
operating practice.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No. The proposed change does not alter the manner in
which safety limits, limiting safety system settings or limiting
conditions for operation are determined. The safety analysis
acceptance criteria are not impacted by this change. Redundant ESFAS
trains are maintained, and diversity with regard to the signals that
provide engineered safety features actuation is also maintained. All
signals credited as primary or secondary, and all operator actions
credited in the accident analysis will remain the same. The proposed
change will not result in unit operation in a configuration outside
the design basis. The calculated impact on risk is insignificant and
meets the acceptance criteria contained in Regulatory Guide 1.174.
The proposed slave relay test interval change will result in a
reduced potential for spurious equipment actuations associated with
testing.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mary O'Reilly, FirstEnergy Nuclear Operating
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH
44308.
NRC Section Chief: Richard J. Laufer.
[[Page 12954]]
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit 1, Lake County, Ohio
Date of amendment request: June 10, 2002.
Description of amendment request: The proposed amendment would
revise Surveillance Requirement (SR) 3.0.3 to extend the delay period,
before entering a Limiting Condition for Operation, following a missed
surveillance. The delay period would be extended from the current limit
of ``* * * up to 24 hours or up to the limit of the specified
Frequency, whichever is less'' to ``* * * up to 24 hours or up to the
limit of the specified Frequency, whichever is greater.'' In addition,
the following requirement would be added to SR 3.0.3: ``A risk
evaluation shall be performed for any Surveillance delayed greater than
24 hours and the risk impact shall be managed.''
The NRC staff issued a notice of opportunity for comment in the
Federal Register on June 14, 2001 (66 FR 32400), on possible amendments
concerning missed surveillances, including a model safety evaluation
and model no significant hazards consideration (NSHC) determination,
using the consolidated line item improvement process. The NRC staff
subsequently issued a notice of availability of the models for
referencing in license amendment applications in the Federal Register
on September 28, 2001 (66 FR 49714). The licensee affirmed the
applicability of the following NSHC determination in its application
dated June 10, 2002.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change relaxes the time allowed to perform a missed
surveillance. The time between surveillances is not an initiator of
any accident previously evaluated. Consequently, the probability of
an accident previously evaluated is not significantly increased. The
equipment being tested is still required to be operable and capable
of performing the accident mitigation functions assumed in the
accident analysis. As a result, the consequences of any accident
previously evaluated are not significantly affected. Any reduction
in confidence that a standby system might fail to perform its safety
function due to a missed surveillance is small and would not, in the
absence of other unrelated failures, lead to an increase in
consequences beyond those estimated by existing analyses. The
addition of a requirement to assess and manage the risk introduced
by the missed surveillance will further minimize possible concerns.
Therefore, this change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or a change in the methods governing normal plant operation. A
missed surveillance will not, in and of itself, introduce new
failure modes or effects and any increased chance that a standby
system might fail to perform its safety function due to a missed
surveillance would not, in the absence of other unrelated failures,
lead to an accident beyond those previously evaluated. The addition
of a requirement to assess and manage the risk introduced by the
missed surveillance will further minimize possible concerns. Thus,
this change does not create the possibility of a new or different
kind of accident from any accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The extended time allowed to perform a missed surveillance does
not result in a significant reduction in the margin of safety. As
supported by the historical data, the likely outcome of any
surveillance is verification that the LCO [Limiting Condition for
Operation] is met. Failure to perform a surveillance within the
prescribed frequency does not cause equipment to become inoperable.
The only effect of the additional time allowed to perform a missed
surveillance on the margin of safety is the extension of the time
until inoperable equipment is discovered to be inoperable by the
missed surveillance. However, given the rare occurrence of
inoperable equipment, and the rare occurrence of a missed
surveillance, a missed surveillance on inoperable equipment would be
very unlikely. This must be balanced against the real risk of
manipulating the plant equipment or condition to perform the missed
surveillance. In addition, parallel trains and alternate equipment
are typically available to perform the safety function of the
equipment not tested. Thus, there is confidence that the equipment
can perform its assumed safety function.
Therefore, this change does not involve a significant reduction
in a margin of safety.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Anthony J. Mendiola.
Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold Energy
Center, Linn County, Iowa
Date of amendment request: February 28, 2003.
Description of amendment request: The proposed amendment would
change the Technical Specifications (TSs) to relocate the numerical
values and curves for the pressure and temperature (P/T) limits for the
reactor coolant system (RCS). The numerical values and curves would be
relocated from the TS to a licensee-controlled document, the Pressure
and Temperature Limits Report (PTLR) pursuant to Nuclear Regulatory
Commission (NRC) Generic Letter (GL) 96-03, ``Relocation of the
Pressure Temperature Limit Curves and Low Temperature Overpressure
Protection System Limits,'' dated January 31, 1996, as modified by NRC
Improved Standard TS, TS Task Force (TSTF) change package number 419,
Revision 0. Specifically, a definition for the PTLR would be added to
TS 1.0, ``Definitions;'' administrative controls for the generation and
reporting requirements associated with the PTLR would be added to TS
5.6, ``Administrative Controls--Reporting Requirements; '' TSs 3.4.9
and 4.4.9 would be modified by removing the numerical values and curve
(Figure 3.4.9-1) for the various P/T limits (which the licensee has
updated using an NRC-approved methodology) and replacing them with a
reference to the PTLR.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The P/T limits are not derived from Design Basis Accident (DBA)
analyses. They are prescribed by the ASME [American Society of
Mechanical Engineers Boiler and Pressure Vessel] Code and 10 CFR
[Part] 50 Appendi[ces] G and H as restrictions on normal operation
to avoid encountering
[[Page 12955]]
pressure, temperature, and temperature rate of change conditions
that might cause undetected flaws to propagate and cause non-ductile
failure of the reactor coolant pressure boundary. Thus, they ensure
that an accident precursor is not likely. Hence, they are included
in the TS as satisfying Criterion 2 of 10 CFR 50.36(c)(2)(ii). The
relocation of the numerical value of these limits to a licensee-
controlled document does not remove the existing TS requirement that
the limits be met. The new TS administrative controls for the PTLR
will ensure that only NRC-approved methods are used to calculate the
actual limits to be applied. Thus, this relocation will not increase
the probability of any accident previously evaluated.
The proposed changes do not alter the design assumptions,
conditions, or configuration of the facility or the manner in which
the facility is operated or maintained. The proposed changes will
not affect any other System, Structure or Component (SSC) designed
for the mitigation of previously analyzed events. The proposed
changes do not affect the source term, containment isolation, or
radiological release assumptions used in evaluating the radiological
consequences of any accident previously evaluated. Thus, the
proposed relocation of the existing numerical values and the updated
figure for the RCS P/T limits based upon an NRC-approved
methodology, to a licensee-controlled document (i.e., the PTLR),
with all the requisite TS restrictions placed upon it by NRC Generic
Letter 96-03, as modified by TSTF-419, Rev. 0, will not increase the
consequences of any previously evaluated accident.
(2) The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed changes do not involve a physical alteration of the
plant (i.e., no new or different type of equipment will be
installed) or a change in the methods governing normal plant
operation. In addition, the changes do not impose any new or
different requirements or eliminate any existing requirements. The
changes do not alter assumptions made in the safety analysis. The
proposed changes are consistent with the safety analysis assumptions
and current plant operating practice. We are merely requesting to
move the existing numerical values and the updated figure for the
RCS P/T limits based upon an NRC-approved methodology, from the TS
to a licensee-controlled document (i.e., the PTLR), with all the
requisite TS restrictions placed upon it by NRC Generic Letter 96-
03, as modified by TSTF-419, Rev. 0.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
(3) The proposed amendment will not involve a significant
reduction in a margin of safety.
The proposed changes do not alter the manner in which Safety
Limits, Limiting Safety System Settings or Limiting Conditions for
Operation are determined. The setpoints at which protective actions
are initiated are not altered by the proposed changes. Sufficient
equipment remains available to actuate upon demand for the purpose
of mitigating an analyzed event. We are merely requesting to move
the existing numerical values and the updated figure for the RCS P/T
limits based upon an NRC-approved methodology, from the TS to a
licensee-controlled document (i.e., the PTLR), with all the
requisite TS restrictions placed upon it by NRC Generic Letter 96-
03, as modified by TSTF-419, Rev. 0. Thus, the proposed changes will
not significantly reduce any margin of safety that currently exists.
Based upon the above, NMC [Nuclear Management Company] has
determined that the proposed amendment will not involve a
significant hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. Alvin Gutterman, Morgan Lewis, 1111
Pennsylvania Avenue NW Washington, DC 20004.
NRC Section Chief: L. Raghavan.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: January 27, 2003.
Description of amendment request: The proposed amendment would make
administrative and editorial changes to the Fort Calhoun Station (FCS)
Technical Specifications (TS) 1.3 Basis (1); 2.7(1)a; 2.7(1)b; 2.7(1)d;
2.7(1)i; 2.7 Basis; 3.0.2; Table 3-5, Item 11; and 3.5(3)ii. The
proposed changes consist primarily of editorial and typographical
changes or corrections.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The correction of typographical errors and clarification of
specifications is not an initiator of any previously evaluated
accident. The frequency or periodicity of performance of those
surveillances affected by this change are not an initiator of any
previously evaluated accident. The proposed changes will not prevent
safety systems from performing their accident mitigation function as
assumed in the safety analysis.
Therefore, this change does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change only affects the technical specifications
and does not involve a physical change to the plant. Modifications
will not be made to existing components nor will any new or
different types of equipment be installed. The proposed change
corrects typographical errors, provides clarification as to
applicable equipment and modifies the frequency of surveillances
performed once per shift from 8 hours to 12 hours. This change will
not alter assumptions made in safety analysis and licensing bases.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change corrects typographical errors, provides
clarification as to applicable equipment, and modifies the frequency
of surveillances performed once per shift from 8 hours to 12 hours.
The decrease in frequency or periodicity of performance of these
surveillances will also permit more efficient and more safely
managed plant operations and can help reduce the risk associated
with changing plant equipment or operating modes in order to obtain
some of these readings.
Therefore, this technical specification change does not involve
a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: James R. Curtiss, Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Stephen Dembek.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: January 27, 2003.
Description of amendment request: The proposed amendment would
delete the allowance to perform the surveillance test of Table 3-2,
Item 20 (Recirculation Actuation Logic Channel Functional Test) under
administrative controls, while components in excess of those allowed by
Conditions a, b, d, and e of Technical Specification 2.3(2) are
inoperable provided they are returned to operable status within one
hour. This allowance was granted in Amendment No. 206 issued April 19,
2002, on an exigent basis and applies only for the remainder of the
current cycle. Omaha
[[Page 12956]]
Public Power District committed to submit a permanent resolution to
this allowance and this license amendment request constitutes this
permanent resolution.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Deleting the requirement to perform the quarterly surveillance
test of Table 3-2, Item 20 (Recirculation Actuation Logic Channel
Functional Test) under administrative controls is acceptable since
the performance of the recirculation actuation logic channel
functional test is not identified as the initiator of any analyzed
event. The proposed change will still require that the surveillance
test be performed and the required ECCS [emergency core cooling
system] systems to be available. This change will not alter
assumptions relative to the mitigation of an accident or transient
event. The performance of this activity has no effect on any
accident scenario. Therefore, the proposed change does not involve a
significant increase in the consequences of an accident previously
evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
This change only removes a short term allowance to utilize
administrative controls in the performance of the recirculation
actuation logic channel functional test. These proposed changes do
not involve a physical alteration of the plant (no new or different
type of equipment will be installed) or change the methods governing
plant operation. The proposed change does not involve any physical
changes to plant systems, structures or components (SSCs) or the
manner in which these SSCs are operated, maintained, modified or
inspected. Therefore, these changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The minimum numbers of ECCS components required by the FCS [Fort
Calhoun Station] accident analyses will remain available. The
proposed change to delete the short term allowance to utilize
administrative controls in the performance of the recirculation
actuation logic channel functional test will not significantly
impact the availability or reliability of the plant's systems or
their ability to respond to plant transients and accidents. The
performance of this activity has no effect on any accident scenario.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: James R. Curtiss, Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Stephen Dembek.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: January 27, 2003.
Description of amendment request: The proposed amendment would
authorize the revision of the Fort Calhoun Station, Unit No. 1 Updated
Safety Analysis Report (USAR). Section 14.16 and Figures 14.16-1
through 14.16-4 of the USAR will be revised to reflect the use of the
GOTHIC, version 7.0, computer code and the results associated with the
updated containment pressure analyses for a loss-of-coolant accident
and main steam line break. In addition, GOTHIC will be used for the
analysis of future plant upgrades associated with containment response
and will be maintained consistent with other NRC-approved Omaha Public
Power District methodologies.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes will not increase the probability or
consequence of any accident based on the following:
The proposed changes to Section 14.16 of the Updated Safety
Analysis Report (USAR) and replacements for Figures 14.16-1 through
14.16-4 is required due to using GOTHIC, version 7.0 and the updated
containment pressure analyses. Demonstrating that containment
pressure is maintained less than the containment design pressure is
required by Fort Calhoun Station (FCS) design basis. Additionally,
the analyses credit all modes of heat transfer defined by Reference
10.5. Therefore, the updated containment pressure analyses using
GOTHIC, version 7.0 is in compliance with FCS design basis. Changes
to the containment pressure analyses for either a loss-of-coolant
accident or main steam line break will be controlled by 10 CFR
50.59. Therefore, the probability or consequence of any accident is
not increased.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed revision does not change any equipment required to
mitigate the consequences of an accident. The continued use of the
same USAR administrative controls prevents the possibility of a new
or different kind of accident. Since the proposed changes do not
involve the addition or modification of equipment nor alter the
design of plant systems, the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated. The changes proposed do not change how design
basis accident events are postulated nor do the changes themselves
initiate a new kind of accident or failure mode with a unique set of
conditions (proposed administrative controls). Therefore, the
proposed change does not create the possibility of a new or
different kind of accident from any previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The use of GOTHIC, version 7.0 is in compliance with FCS design
basis. Additionally, GOTHIC has been benchmarked to the current
analysis of record for a loss-of-coolant accident and main steam
line break using the NRC approved computer code CONTRANS. These
benchmark models demonstrate that GOTHIC provides similar results to
CONTRANS. Future updates of the containment pressure analyses will
be conducted under the 10 CFR 50.59 process. The analyses will
credit all available modes of heat transfer defined by Reference
10.5. Additionally, the main steam line break containment evaluation
model considers the leakage past the broken steam generator main
feed isolation valve of 2.45% of full power flow or approximately
195 gpm. Therefore, the proposed changes do not involve a
significant reduction to the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: James R. Curtiss, Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Stephen Dembek.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: January 27, 2003.
Description of amendment request: The proposed amendment revises
Technical Specifications (TS) 2.1.6, 3.2
[[Page 12957]]
(Table 3-5), and 5.9.1c. For TS 2.1.6(1), Omaha Public Power District
(OPPD) has proposed to increase the ``as-found'' pressurizer safety
valve (PSV) lift setting tolerance band of +/-1% to +1%/-3% to allow
for normal setpoint variance for Modes 1 and 2. The Basis of TS 2.1.6
will be revised to clarify that the PSVs are still operable and capable
of performing their safety function with the wider tolerance band. The
remaining revisions to TS 2.1.6 are administrative in nature to change
defined terms to upper case text. OPPD has also proposed to revise (1)
item 3 in Table 3-5 of TS 3.2 to require an ``as-left'' PSV lift
setting tolerance band of +/-1%, and (2) TS 5.9.1c to remove the
requirement to provide a statement in the Monthly Operating Report
(MOR) concerning failures or challenges to power operated relief valves
(PORV) or safety valves.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The design basis event for RCS over-pressure protection is the
Loss of Load accident. The Loss of Load event was previously
evaluated assuming the PSVs lift up to 6% above their setpoint.
While the proposed amendment widens the tolerance band for installed
PSVs, only the lower end of the band is changed; therefore, there is
no adverse affect on the over-pressure protection analysis.
The proposed amendment does not change the tolerance band
currently required at the conclusion of PSV surveillance testing
each refueling outage. As with the current specification, the PSVs
will continue to be set to within a tolerance band of +/- 1% using
ASME Code test methods. As a result, the anticipated performance of
the valves over the course of the subsequent operating cycle is not
changed. In other words, the potential for setpoint variance exists
regardless of whether the TSs are changed. The PSVs will begin each
operating cycle after having been set to open within a lift setting
tolerance band of +/- 1%. Therefore, the probability or consequences
of potential setpoint variance during an operating cycle does not
change. The remaining changes provide supporting statements for the
wider PSV lift setting tolerance band in the Basis of TS 2.1.6, are
administrative in nature, or are in accordance with GL 97-02.
The changes in the case of the defined terms and elimination of
the TS 5.9.1c Monthly Operating Report concerning failures or
challenges to PORVs or safety valves are administrative changes
which do not affect the initiator of an event or prevent safety
systems from performing their accident mitigation functions as
assumed in the safety analysis.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Widening the lift setting tolerance band for installed PSVs does
not create the possibility of a new or different type of accident
from any previously evaluated.
The accident analyses address the lift setting tolerance band of
the PSVs, and the proposed tolerance band does not adversely affect
the over-pressure protection function and will not compromise RCS
integrity during power operation. No physical changes to the plant
are involved.
The proposed amendment does not change the tolerance band that
must be met at the conclusion of PSV surveillance testing each
refueling outage. As with the current Technical Specifications, the
PSVs will continue to be set at a tolerance band of +/- 1% using
ASME Code test methods. As a result, the anticipated performance of
the valves over the course of the subsequent operating cycle is not
changed. The remaining changes provide supporting statements for the
wider PSV lift setting tolerance band in the Basis of TS 2.1.6, are
administrative in nature, or are in accordance with GL 97-02 and
thus do not create the possibility of a new or different type of
accident from any previously evaluated.
The changes in the case of the defined terms and elimination of
the TS 5.9.1c Monthly Operating Report concerning failures or
challenges to PORVs or safety valves are administrative changes
which only affect the technical specifications and do not involve a
physical change to the plant. Therefore these changes do not alter
assumptions made in the safety analysis and licensing basis.
3. The proposed change does not involve a significant reduction
in a margin of safety.
Widening the lift setting tolerance band for installed PSVs does
not involve a significant reduction in a margin of safety. The
tolerance band of the PSVs is addressed in the accident analyses,
and the proposed tolerance band does not adversely affect the over-
pressure protection analysis. No physical changes to the plant are
involved.
The proposed amendment does not change the tolerance band that
must be met at the conclusion of PSV surveillance testing each
refueling outage. As with the current Technical Specifications, the
PSVs will continue to be set to a tolerance band of +/- 1% using
ASME Code test methods. As a result, the anticipated performance of
the valves over the course of the subsequent operating cycle is not
changed. The remaining changes provide supporting statements for the
wider PSV lift setting tolerance band in the Basis of TS 2.1.6, are
administrative in nature, or are in accordance with GL 97-02.
The changes in the case of the defined terms and elimination of
the TS 5.9.1c Monthly Operating Report concerning failures or
challenges to PORVs or safety valves are administrative changes
which only affect the technical specifications and reporting
frequency.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: James R. Curtiss, Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Stephen Dembek.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, Alabama
Date of amendment request: February 19, 2003.
Description of amendment request: The proposed amendments delete
requirements from the technical specifications (TS) and other elements
of the licensing bases to maintain a Post Accident Sampling System
(PASS). Licensees were generally required to implement PASS upgrades as
described in NUREG-0737, ``Clarification of TMI [Three Mile Island]
Action Plan Requirements,'' and Regulatory Guide 1.97,
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess
Plant and Environs Conditions During and Following an Accident.''
Implementation of these upgrades was an outcome of the lessons learned
from the accident that occurred at TMI Unit 2. Requirements related to
PASS were imposed by Order for many facilities and were added to or
included in the TS for nuclear power reactors currently licensed to
operate. Lessons learned and improvements implemented over the last 20
years have shown that the information obtained from PASS can be readily
obtained through other means or is of little use in the assessment and
mitigation of accident conditions.
The changes are based on NRC-approved Technical Specification Task
Force (TSTF) Standard Technical Specification Change Traveler, TSTF-
413, ``Elimination of Requirements for a Post Accident Sampling System
(PASS).'' The NRC staff issued a notice of opportunity for comment in
the Federal Register on December 27, 2001 (66 FR 66949), on possible
amendments concerning TSTF-413, including a model safety evaluation and
model no
[[Page 12958]]
significant hazards consideration (NSHC) determination, using the
consolidated line item improvement process. The NRC staff subsequently
issued a notice of availability of the models for referencing in
license amendment applications in the Federal Register on March 20,
2002 (67 FR 13027). The licensee affirmed the applicability of the
following NSHC determination in its application dated February 19,
2003.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The PASS was originally designed to perform many sampling and
analysis functions. These functions were designed and intended to be
used in post accident situations and were put into place as a result
of the TMI-2 accident. The specific intent of the PASS was to
provide a system that has the capability to obtain and analyze
samples of plant fluids containing potentially high levels of
radioactivity, without exceeding plant personnel radiation exposure
limits. Analytical results of these samples would be used largely
for verification purposes in aiding the plant staff in assessing the
extent of core damage and subsequent offsite radiological dose
projections. The system was not intended to and does not serve a
function for preventing accidents and its elimination would not
affect the probability of accidents previously evaluated.
In the 20 years since the TMI-2 accident and the consequential
promulgation of post accident sampling requirements, operating
experience has demonstrated that a PASS provides little actual
benefit to post accident mitigation. Past experience has indicated
that there exists in-plant instrumentation and methodologies
available in lieu of a PASS for collecting and assimilating
information needed to assess core damage following an accident.
Furthermore, the implementation of Severe Accident Management
Guidance (SAMG) emphasizes accident management strategies based on
in-plant instruments. These strategies provide guidance to the plant
staff for mitigation and recovery from a severe accident. Based on
current severe accident management strategies and guidelines, it is
determined that the PASS provides little benefit to the plant staff
in coping with an accident.
The regulatory requirements for the PASS can be eliminated
without degrading the plant emergency response. The emergency
response, in this sense, refers to the methodologies used in
ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities. The
elimination of the PASS will not prevent an accident management
strategy that meets the initial intent of the post-TMI-2 accident
guidance through the use of the SAMGs, the emergency plan (EP), the
emergency operating procedures (EOP), and site survey monitoring
that support modification of emergency plan protective action
recommendations (PARs).
Therefore, the elimination of PASS requirements from Technical
Specifications (TS) (and other elements of the licensing bases) does
not involve a significant increase in the consequences of any
accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The elimination of PASS related requirements will not result in
any failure mode not previously analyzed. The PASS was intended to
allow for verification of the extent of reactor core damage and also
to provide an input to offsite dose projection calculations. The
PASS is not considered an accident precursor, nor does its existence
or elimination have any adverse impact on the pre-accident state of
the reactor core or post accident confinement of radioisotopes
within the containment building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The elimination of the PASS, in light of existing plant
equipment, instrumentation, procedures, and programs that provide
effective mitigation of and recovery from reactor accidents, results
in a neutral impact to the margin of safety. Methodologies that are
not reliant on PASS are designed to provide rapid assessment of
current reactor core conditions and the direction of degradation
while effectively responding to the event in order to mitigate the
consequences of the accident. The use of a PASS is redundant and
does not provide quick recognition of core events or rapid response
to events in progress. The intent of the requirements established as
a result of the TMI-2 accident can be adequately met without
reliance on a PASS.
Therefore, this change does not involve a significant reduction
in the margin of safety.
The NRC staff proposes to determine that the amendment requests
involve no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Section Chief: Allen G. Howe.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of amendment request: December 13, 2003.
Description of amendment request: The proposed amendment would
allow the use of Westinghouse leak-limiting Alloy 800 sleeves to repair
defective steam generator tubes as an alternative to plugging the tube.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration in accordance with the three standards set forth in 10
CFR 50.92(c), which are presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
No. The Westinghouse Alloy 800 leak-limiting repair sleeves are
designed using the applicable American Society of Mechanical
Engineers (ASME) Boiler and Pressure Vessel Code and, therefore,
meet the design objectives of the original steam generator tubing.
The applied stresses and fatigue usage for the repair sleeves are
bounded by the limits established in the ASME Code. Mechanical
testing has shown that the structural strength of repair sleeves
under normal, upset, emergency, and faulted conditions provides
margin to the acceptance limits. These acceptance limits bound the
most limiting (three times normal operating pressure differential)
burst margin recommended by NRC's Regulatory Guide 1.121, ``Bases
for Plugging Degraded PWR Steam Generator Tubes.'' Burst testing of
sleeve/tube assemblies has demonstrated that no unacceptable levels
of primary-to-secondary leakage are expected during any plant
condition.
The Alloy 800 repair sleeve depth-based structural limit is
determined using the NRC guidance and the pressure stress equation
of ASME Code, Section III with additional margin added to account
for configuration of long axial cracks. A bounding detection
threshold value has been conservatively identified and statistically
established to account for growth and determine the repair sleeve/
tube assembly plugging limit. A sleeved tube is plugged on detection
of degradation in the sleeve/tube assembly.
Evaluation of the repaired steam generator tube testing and
analysis indicates no detrimental effects on the sleeve or sleeved
tube assembly from reactor system flow, primary or secondary coolant
chemistries, thermal conditions or transients, or pressure
conditions as may be experienced at Watts Bar Unit 1. Corrosion
testing and historical performance of sleeve/tube assemblies
indicates no evidence of sleeve or tube corrosion considered
detrimental under anticipated service conditions.
The implementation of the proposed amendment has no significant
effect on either the configuration of the plant or the manner in
which it is operated. The consequences of a hypothetical failure of
the sleeve/tube assembly is bounded by the current steam generator
tube rupture (SGTR) analysis described in Watts Bar Unit 1 Updated
Final Safety Analysis Report. Due to the slight reduction in
diameter caused by the sleeve wall thickness, primary coolant
release rates
[[Page 12959]]
would be slightly less than assumed for the steam generator tube
rupture analysis and; therefore, would result in lower total primary
fluid mass release to the secondary system. A main steam line break
or feedwater line break will not cause a SGTR since the sleeves are
analyzed for a maximum accident differential pressure greater that
that predicted in the Watts Bar Unit 1 safety analysis. The minimal
repair sleeve/tube assembly leakage that could occur during plant
operation is well within the Technical Specification leakage limits
when grouped with current alternate plugging criteria calculated
leakage values.
Therefore, TVA has concluded that the proposed change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
No. The Alloy 800 leak-limiting repair sleeves are designed
using the applicable ASME Code as guidance; therefore, it meets the
objectives of the original steam generator tubing. As a result, the
functions of the steam generators will not be significantly affected
by the installation of the proposed sleeve. The proposed repair
sleeves do not interact with any other plant systems. Any accident
as a result of potential tube or sleeve degradation in the repaired
portion of the tube is bounded by the existing SGTR accident
analysis. The continued integrity of the installed sleeve/tube
assembly is periodically verified by the Technical Specification
requirements and the sleeved tube plugged on detection of
degradation.
The implementation of the proposed amendment has no significant
effect on either the configuration of the plant, or the manner in
which it is operated. Therefore, TVA concludes that this proposed
change does not create the possibility of a new or different kind of
accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
No. The repair of degraded steam generator tubes with Alloy 800
leak-limiting repair sleeves restores the structural integrity of
the degraded tube under normal operating and postulated accident
conditions and thereby maintains current core cooling margin as
opposed to plugging the tube and taking it out of service. The
design safety factors utilized for the repair sleeves are consistent
with the safety factors in the ASME Boiler and Pressure Vessel Code
used in the original steam generator design. The portions of the
installed sleeve/tube assembly that represent the reactor coolant
pressure boundary can be monitored for the initiation of sleeve/tube
wall degradation and affected tube plugged on detection. Use of the
previously identified design criteria and design verification
testing assures that the margin to safety is not significantly
different from the original steam generator tubes.
Therefore, TVA concludes that the proposed change does not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
NRC Section Chief: Allen G. Howe.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of amendment request: December 13, 2002.
Description of amendment request: The proposed amendment would
revise the Watts Bar Nuclear Plant, Unit 1, Technical Specifications to
add two new Sections, 3.7.16, ``Shutdown Board Room Air Conditioning
System,'' and 3.7.17, ``Elevation 772.0 480 Volt Board Room Air
Conditioning Systems.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration in accordance with the three standards set forth in 10
CFR 50.92(c), which are presented below:
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
[No.] The proposed revision to the [Watts Bar Nuclear Plant] TS
will provide formalized operational guidance for coping with partial
or complete unavailability of SDBR [shutdown board room] and 480V
board room air conditioning (AC) equipment for limited periods of
time. The change does not impact the frequency of an accident
because failure of either the SDBR or the 480V board room AC systems
is not an initiator of any accident scenario. The change does not
modify any plant hardware including the air conditioning systems,
and none of their automatic control features or redundant systems
currently credited in failure analyses are being deleted, modified,
or otherwise replaced by operator actions as a result of the
proposed change.
The proposed TS revision changes current plant operating
practice and WBN Final Safety Analysis Report (FSAR) assumptions by
allowing continued power operation with both trains of SDBR air
conditioning concurrently inoperable and two 480V board room AC
systems of the same unit to be concurrently inoperable for a limited
duration, up to 12 hours. This condition is acceptable based on the
low probability of the occurrence of postulated accidents resulting
in core damage concurrent with multiple inoperable systems or trains
of cooling equipment during this timeframe, and based on analyses
which demonstrate that peak temperatures in each room served by
these systems remain below mild environment temperature limits
during this time period. Consequently, there is no significant
adverse impact on the ability of required safety-related electrical
equipment to continue to operate and perform their required
functions, during both normal operation and during design basis
events. Therefore, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
[No.] The proposed change does not modify any plant hardware
including the subject air conditioning systems. The change provides
specific operational guidance for coping with partial or complete
unavailability of SDBR and 480V board room air conditioning
equipment. No new accident or event initiators are created by
allowing multiple air conditioning systems to be unavailable for the
limited time period of 12 hours. The supported electrical equipment
remains capable of performing its intended function both during
normal operations and post accident. Therefore, the proposed changes
do not create the possibility of a new or different kind of accident
from any accident previously evaluated.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
[No.] The proposed TS revision changes current FSAR assumptions
by allowing continued power operation with both trains of SDBR air
conditioning concurrently inoperable and allowing two 480V board
room air conditioning systems of the same unit to be inoperable for
a limited duration, up to 12 hours. This condition does not
significantly reduce the margin of safety due to the low probability
of the occurrence of a postulated accident resulting in core damage
concurrent with multiple inoperable systems or trains of cooling
equipment during the limited time period. In addition, transient
temperature analyses demonstrate that peak temperatures in each room
served by these systems remain below mild environment temperature
limits for a period of 24 hours assuming a complete loss of air
conditioning to all rooms served by the SDBR and 480V board room AC
systems concurrently. The analysis is bounding for normal
operational conditions. Consequently, there is no significant
adverse impact on the ability of required safety-related electrical
equipment to continue to operate and perform their required
functions during both normal operation and during design basis
events. Therefore, the proposed change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
[[Page 12960]]
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
NRC Section Chief: Allen G. Howe.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management Systems (ADAMS) Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the NRC Public Document Room (PDR) Reference staff at 1-800-
397-4209, 301-415-4737 or by e-mail to [email protected].
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of application for amendments: September 20, 2002.
Brief description of amendments: These amendments adopt the generic
changes approved by Technical Specification Task Force (TSTF) change
travelers TSTF-349, Revision 1, and TSTF-361, Revision 2, for NUREG-
1430, Revision 1, ``Standard Technical Specifications, Babcock and
Wilcox Plants,'' dated April 1995, and incorporated into NUREG-1430,
Revision 2, dated June 2001. Specifically, Section 3.9.5, ``Shutdown
Cooling (SDC) and Coolant Circulation--Low Water Level,'' is revised to
add two notes to allow operational changes in the shutdown cooling
system.
Date of issuance: February 25, 2003.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 256 and 233.
Renewed Facility Operating License Nos. DPR-53 and DPR-69:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: October 29, 2002 (67 FR
66007).
The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated February 25, 2003.
No significant hazards consideration comments received: No.
Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power
Station, Unit No. 2, New London County, Connecticut
Date of application for amendment: February 5, 2002, as
supplemented January, 14, 2003.
Brief description of amendment: The amendment revises the
surveillance requirements associated with the Containment Isolation
Valves (CIVs), Reactor Building Closed Cooling Water (RBCCW) System,
and Service Water (SW) System to remove redundant testing requirements
that are already addressed by the Inservice Testing Program. Additional
changes remove the post maintenance testing requirements associated
with the CIVs, revise the wording of the RBCCW and SW Systems Limiting
Conditions for Operation, and increase the allowed outage times for the
RBCCW and SW Systems.
Date of issuance: February 13, 2003.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No.: 273.
Facility Operating License No. DPR-65: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 16, 2002 (67 FR
18644). The January 14, 2003, letter provided clarifying information
that did not change the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 13, 2003.
No significant hazards consideration comments received: No.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: May 14, 2002, as supplemented by letter
dated December 20, 2002.
Brief description of amendment: The amendment changes
administrative Technical Specification 5.5.13 regarding the Containment
Integrated Leak Rate Testing (ILRT) to allow a one-time extension of
the interval (to 15 years) for performance of the next ILRT.
Date of issuance: March 5, 2003.
Effective date: As of the date of issuance and shall be implemented
30 days from the date of issuance.
Amendment No.: 131.
Facility Operating License No. NPF-47: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 25, 2002 (67 FR
42823).
The December 20, 2002, supplemental letter provided clarifying
information that did not change the scope of the original Federal
Register notice or the original no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 5, 2003.
No significant hazards consideration comments received: No.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon,
Vermont
Date of application for amendment: December 10, 2002, as
supplemented on January 20, 2003.
Brief description of amendment: The Technical Specification (TS)
amendment request changes the diesel fuel specification to a more
current revision in TS 4.10.C. The changes also
[[Page 12961]]
make administrative revisions to reflect generic position titles in TS
6.0; correct page numbers and titles in the Table of Contents; and to
delete the General Table of Contents. Bases pages were also revised to
reflect the fuel specification revision, as well as to make
administrative changes to provide clarity and correct a misspelling.
Date of Issuance: February 27, 2003.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 214.
Facility Operating License No. DPR-28: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 21, 2003 (68 FR
2802).
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated February 27, 2003.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of application for amendments: September 27, 2002.
Brief description of amendments: The amendments change Appendix B,
``Environmental Protection Plan,'' of the licensee by removing a
parenthetical reference to a superseded section of 10 CFR part 51.
Date of issuance: February 20, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 157/143
Facility Operating License Nos. NPF-11 and NPF-18: The amendments
revised the Environmental Protection Plan.
Date of initial notice in Federal Register: October 29, 2002 (67 FR
66009).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 20, 2003.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2, Beaver County,
Pennsylvania
Date of application for amendments: August 7, 2002.
Brief description of amendments: The amendments: (1) Revised the
surveillance frequency for air or smoke flow testing of containment
spray nozzles, as specified in surveillance requirements (SRs)
4.6.2.1.d and 4.6.2.2.f, from, ``once per 10 years,'' to, ``following
maintenance which results in the potential for nozzle blockage as
determined by engineering evaluation;'' (2) allowed the use of a visual
examination in lieu of an air or smoke flow test; (3) relocated the SR
4.6.2.2.e.3 criteria for the river/service water flow rate through the
recirculation spray system heat exchangers to the Updated Final Safety
Analysis Report; and (4) made minor clarifying changes to the text in
TS 3.3.1.1.
Date of issuance: February 24, 2003.
Effective date: As of date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 252 and 132.
Facility Operating License Nos. DPR-66 and NPF-73: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 15, 2002 (67 FR
63694).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 24, 2003.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2, Beaver County,
Pennsylvania
Date of application for amendments: March 14, 2002.
Brief description of amendments: The amendments revised the
Technical Specifications (TSs) by extending the allowed outage time
(AOT), or completion time, associated with an inoperable emergency core
cooling system (ECCS) accumulator. In addition to the AOT extension,
other changes were incorporated to make the ECCS TSs consistent with
NUREG-1431, ``Standard Technical Specifications--Westinghouse Plants.''
Format and editorial changes were included as necessary to facilitate
the revision of the TS text to conform to the current TS page format.
Date of issuance: February 25, 2003.
Effective date: As of date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 253 and 133.
Facility Operating License Nos. DPR-66 and NPF-73: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 30, 2002 (67 FR
21289).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 25, 2003.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2, Beaver County,
Pennsylvania
Date of application for amendments: October 31, 2002, as
supplemented by letters dated December 2, 2002, and January 24, 2003.
Brief description of amendments: The amendments revised the
Technical Specifications to allow extending the Type A containment
integrated leak rate test interval from 10 years to 15 years on a one-
time basis.
Date of issuance: March 5, 2003.
Effective date: As of date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 254 and 134.
Facility Operating License Nos. DPR-66 and NPF-73: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 10, 2002 (67
FR 75877).
The December 2, 2002, and January 24, 2003, supplemental letters
did not change the initial no significant hazards consideration
determination or expand the amendment beyond the scope of the initial
notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 5, 2003.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit 1, Lake County, Ohio
Date of application for amendment: December 9, 2002.
Brief description of amendment: Pursuant to 10 CFR 50.67, this
amendment approves the use of Alternative Source Term radiological
calculations to update the design bases analysis for the Fuel Handling
Accident as described in the Updated Safety Analysis Report. Regulatory
Guide 1.183, ``Alternative Radiological Source Terms for Evaluating
Design-Basis Accidents at Nuclear Power Reactors,'' was used in the
application.
Date of issuance: March 4, 2003.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 122.
Facility Operating License No. NPF-58: This amendment revised the
Updated Safety Analysis Report.
Date of initial notice in Federal Register: January 7, 2003 (68 FR
804).
[[Page 12962]]
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 4, 2003.
No significant hazards consideration comments received: No.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant, Units 3 and 4, Miami-Dade County, Florida
Date of application for amendments: August 16, 2002.
Brief description of amendments: The proposed amendments modified
Technical Specification (TS) Surveillance Requirement Section 4.0.3 to
extend the delay time for completion of a missed surveillance to 24
hours or up to the surveillance frequency, whichever is greater.
Additionally the proposed change would add a TS Bases Control Program.
Date of issuance: March 3, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos: 222 and 217.
Facility Operating License Nos. DPR-31 and DPR-41: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 24, 2002 (67
FR 78521).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 3, 2003.
No significant hazards consideration comments received: No.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant, Units 3 and 4, Miami-Dade County, Florida
Date of application for amendments: October 21, 2002, as
supplemented by letters dated February 11, 2003, and March 3, 2003.
Brief description of amendments: The amendments will reduce the
minimum time required for reactor subcriticality prior to removing
irradiated fuel from the reactor vessel from 100 hours to 72 hours, as
specified in Technical Specification 3/4.9.3 ``Refueling Operations,
Decay Time.''
Date of issuance: March 4, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos: 223 and 218.
Facility Operating License Nos. DPR-31 and DPR-41: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 12, 2002 (67
FR 68738).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 4, 2003.
No significant hazards consideration comments received: No.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of application for amendments: April 11, 2002, as supplemented
November 11, 2002.
Brief description of amendments: The amendments would revise the
Surveillance Requirements for containment leakage rate testing in
Technical Specification 4.6.1.2 to allow a one-time extension of the
interval between integrated leakage rate tests from 10 to 15 years.
Date of issuance: February 25, 2003.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment Nos.: 274 and 254.
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 14, 2002 (67 FR
34488).
The supplemental letter contained clarifying information and did
not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 25, 2003.
No significant hazards consideration comments received: No.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: February 28, 2001, as supplemented by
letters dated February 26, September 13 and 27, and November 25, 2002
(2).
Brief description of amendment: The amendment consists of changes
to the design-basis accidents dose assessment methodology and Operating
License Condition 2.C.(6).
Date of issuance: February 21, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment No.: 196.
Facility Operating License No. DPR-46: Amendment revised the final
safety analysis report and Operating License Condition 2.C.(6).
Date of initial notice in Federal Register: September 19, 2001 (66
FR 48289).
The supplemental letters provided clarifying information that was
within the scope of the original Federal Register notice (66 FR 48289)
and did not change the initial no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 21, 2003.
No significant hazards consideration comments received: No.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile
Point Nuclear Station, Unit 2, Oswego County, New York
Date of application for amendment: February 3, 2003.
Brief description of amendment: The amendment changed Technical
Specifications Surveillance Requirement 3.6.1.7.2 for suppression
chamber-to-drywell vacuum breaker 2ISC*RV36B to allow an exception to
the periodic functional testing requirements for the remainder of Cycle
9.
Date of issuance: February 21, 2003.
Effective date: As of the date of issuance to be implemented within
7 days.
Amendment No.: 108.
Facility Operating License No. NPF-69: Amendment revises the
Technical Specifications. Public comments requested as to proposed no
significant hazards consideration: Yes. The Nuclear Regulatory
Commission published a public notice of the proposed amendment, issued
a proposed finding of no significant hazards consideration and
requested that any comments on the proposed no significant hazards
consideration be provided to the staff by the close of business on
February 20, 2003. The notice was published in the Syracuse, NY, The
Post-Standard, on February 11, 2003.
No significant hazards consideration comments received: No.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, consultation with the State of New York, and
final no significant hazards consideration determination are contained
in a Safety Evaluation dated February 21, 2003.
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of application for amendment: April 22, 2002, as supplemented
September 16, 2002.
Brief description of amendment: The amendment changes the Technical
Specifications by revising the curves for minimum pressure-temperature
for the reactor pressure vessel. The P-T curves addressed by this
amendment were
[[Page 12963]]
developed in accordance with (1) the 1989 edition of the American
Society of Mechanical Engineers (ASME) Code, section Xl, appendix G,
(2) 10 CFR part 50, appendix G, and (3) ASME Code Case N-640,
``Alternative Reference Fracture Toughness for Development of P-T Limit
Curves.''
Date of issuance: February 24, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 133.
Facility Operating License No. DPR-22. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 3, 2002 (67
FR 56323).
The September 16, 2002, supplemental letter provided additional
clarifying information that was within the scope of the original
application, did not change the NRC staff's initial no significant
hazards consideration determination, and did not expand the scope of
the original Federal Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 24, 2003.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-255, Palisades Plant,
Van Buren County, Michigan
Date of application for amendment: March 1, 2002, as supplemented
November 7, 2002.
Brief description of amendment: The amendment revises the testing
frequency for the containment spray nozzles specified in Technical
Specification Surveillance Requirement 3.6.6.9. The testing frequency
for the containment spray nozzles is changed from 10 years to
``following maintenance which could result in nozzle blockage.''
Date of issuance: February 24, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 211.
Facility Operating License No. DPR-20. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 15, 2002 (67 FR
63696).
The November 7, 2002, supplemental letter provided additional
clarifying information that was within the scope of the original
application, did not change the NRC staff's initial no significant
hazards consideration determination, and did not expand the scope of
the original Federal Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 24, 2003.
No significant hazards consideration comments received: No.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: October 8, 2002.
Brief description of amendment: The amendment relocates the
requirements of TS 3.5(5) for testing prestressed concrete containment
tendons to the Fort Calhoun Station, Unit No. 1 Updated Safety Analysis
Report. The amendment adds the requirement for a Containment Tendon
Testing Program (TS 5.21) consistent with that presented in Section 5.5
of NUREG-1432, ``Improved Standard Technical Specification (ITS) for
Combustion Engineering Plants.''
Date of issuance: February 26, 2003.
Effective date: February 26, 2003, and shall be implemented within
120 days from the date of issuance, including the incorporation of the
containment tendons testing requirements into the Updated Safety
Analysis Report.
Amendment No.: 216.
Facility Operating License No. DPR-40: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 12, 2002 (67
FR 68741).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 26, 2003.
No significant hazards consideration comments received: No.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania
Date of application for amendments: October 16, 2001, as
supplemented August 23, 2002, November 8, 2002, and January 20, 2003.
Brief description of amendments: These amendments revised the
technical specifications (TSs) to incorporate seven industry-proposed
Technical Specification Task Force changes (TSTFs) made to NUREG-1433,
Revision 1, ``Standard Technical Specifications for General Electric
Plants (BWR/4),'' that have been approved by the Nuclear Regulatory
Commission.
Date of issuance: February 25, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 209 and 183.
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 12, 2001 (66
FR 64300). The supplements dated August 23, 2002, November 8, 2002, and
January 20, 2003 provided additional information that clarified the
application, did not expand the scope of the application as originally
noticed, and did not change the staff's original proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 25, 2003.
No significant hazards consideration comments received: No.
PPL Susquehanna, LLC, Docket No. 50-388, Susquehanna Steam Electric
Station, Unit 2, Luzerne County, Pennsylvania
Date of application for amendments: July 17, 2002, as supplemented
by letters dated October 30, 2002, December 18, 2002, and January 28,
2003.
Brief description of amendments: The amendment revised the values
of the Safety Limit for Minimum Critical Power Ratio in the Unit 2
Technical Specifications (TSs) 2.1.1.2, clarified fuel design features
in TS 4.2.1, and updated the references used to determine the core
operating limits in TS 5.6.5.b.
Date of issuance: March 4, 2003.
Effective date: As of the date of issuance and shall be implemented
upon startup following the Susquehanna Steam Electric Station, Unit 2
eleventh refueling and inspection outage.
Amendment Nos.: 184.
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 20, 2002 (67 FR
53988).
The supplements dated October 30, 2002, December 18, 2002, and
January 28, 2003, provided additional information that clarified the
application, did not expand the scope of the application as originally
noticed, and did not change the staff's original proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 4, 2003.
No significant hazards consideration comments received: No.
[[Page 12964]]
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of application for amendments: December 2, 2002.
Brief description of amendments: The amendments revised Technical
Specification Surveillance Requirement 3.6.4.1.2 to require that only
one access door in each opening of the secondary containment be closed.
Date of issuance: February 28, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 236/178.
Renewed Facility Operating License Nos. DPR-57 and NPF-5:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: January 7, 2003 (68 FR
812).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 28, 2003.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-260, Browns Ferry Nuclear
Plant, Unit 2, Limestone County, Alabama
Date of application for amendments: October 25, 2002, as
supplemented December 20, 2002, and February 11 and 21, 2003.
Description of amendment request: The amendment updated the values
of the Safety Limit Minimum Critical Power Ratio in Technical
Specification 2.1.1.2 for Cycle 13 operation.
Date of issuance: February 28, 2003.
Effective date: Date of issuance, to be implemented within 60 days.
Amendment No.: 280.
Facility Operating License No. DPR-52: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 10, 2002 (67
FR 75885). The supplemental letters provided clarifying information
that did not change the initial proposed no significant hazards
consideration determination or expand the scope of the original
request.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 28, 2003.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-327, Sequoyah Nuclear Plant,
Unit 1, Hamilton County, Tennessee
Date of application for amendment: March 29, 2002, as supplemented
on October 10, 2002.
Brief description of amendment: The proposed amendment deletes
several of the Unit 1 Technical Specification (TS) Surveillance
Requirements (SR) contained in TS 3/4.4.5, ``Steam Generators'' (SGs),
associated with the voltage-based SG alternative repair criteria. In
addition the proposed changes would delete License Condition 2.C.9.d
which references commitment letters associated with SG inspection
activities.
Date of issuance: March 4, 2003.
Effective date: As of the date of issuance and shall be implemented
during the 2003 Cycle 12 Refueling Outage.
Amendment No.: 282.
Facility Operating License No. DPR-77: Amendment revises the TSs.
Date of initial notice in Federal Register: August 6, 2002 (67 FR
50960). An October 10, 2002 submittal revised some of the information,
so a revised notice was published October 29, 2002 (67 FR 66014).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 4, 2003.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 10th day of March, 2003.
For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 03-6286 Filed 3-17-03; 8:45 am]
BILLING CODE 7590-01-P